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Materials Science & Technology
The objectives of MSTD are: promote the advancement of materials science in Nuclear Science Technology; support the multidisciplines which constitute it; encourage research by providing a forum for the presentation, exchange, and documentation of relevant information; promote the interaction and communication among its members; and recognize and reward its members for significant contributions to the field of materials science in nuclear technology.
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Conference on Nuclear Training and Education: A Biennial International Forum (CONTE 2025)
February 3–6, 2025
Amelia Island, FL|Omni Amelia Island Resort
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The Standards Committee is responsible for the development and maintenance of voluntary consensus standards that address the design, analysis, and operation of components, systems, and facilities related to the application of nuclear science and technology. Find out What’s New, check out the Standards Store, or Get Involved today!
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Nuclear Science and Engineering
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Nuclear Technology
Fusion Science and Technology
Latest News
A more open future for nuclear research
A growing number of institutional, national, and funder mandates are requiring researchers to make their published work immediately publicly accessible, through either open repositories or open access (OA) publications. In addition, both private and public funders are developing policies, such as those from the Office of Science and Technology Policy and the European Commission, that ask researchers to make publicly available at the time of publication as much of their underlying data and other materials as possible. These, combined with movement in the scientific community toward embracing open science principles (seen, for example, in the dramatic rise of preprint servers like arXiv), demonstrate a need for a different kind of publishing outlet.
H. Takenaga, H. Kubo, S. Higashijima, N. Asakura, T. Sugie, S. Konoshima, K. Shimizu, T. Nakano, K. Itami, A. Sakasai, H. Tamai, S. Sakurai, Y. Miura, N. Hosogane, M. Shimada
Fusion Science and Technology | Volume 42 | Number 2 | September-November 2002 | Pages 327-356
Technical Paper | doi.org/10.13182/FST02-A232
Articles are hosted by Taylor and Francis Online.
Heat and particle control has been studied under the reactor-relevant high-power heating in the large tokamak of JT-60U with an open divertor and progressively a W-shaped pumped divertor. Heat and particle control is crucial for reduction in heat load onto the divertor plates, control of density in the main plasma, effective exhaust of helium ash, and reduction in impurity contamination. For the reduction of heat load, radiative divertor concept was developed based on understanding of heat and particle transport in scrape-off layer and divertor plasmas, which contributed to establishment of divertor concept in ITER. With argon injection, the total radiation loss power reached up to 80% of the net heating power with high confinement of HHy2 ~ 1, where HHy2 is a confinement enhancement factor over the IPB98(y,2) ELMy H-mode scaling, at high density of 80% of the Greenwald density in the ELMy H-mode plasma. For the density control, the dependence of particle confinement on plasma parameters was systematically studied with two confinement times for center- and edge-fueled particles, which enabled discussion of density controllability. Core fueling using a high-field-side pellet injection extended the operation range of high confinement (HHy2 ~ 1) from 60 to 70% of the Greenwald density in the high p ELMy H-mode plasma. Efficient helium ash exhaust of He*/E = 2.8 was demonstrated in the ELMy H-mode plasma with the pumping from the private flux region, which is the same pumping geometry as that in ITER design. Reduction in Zeff by puff-and-pump scheme was demonstrated, and chemical sputtering yields were estimated with the consideration of not only methane but also heavier hydrocarbons. Their sputtering yields showed strong dependence on the wall temperature and weak dependence on the particle flux. The measured profiles of C II and C IV line intensities were well reproduced by the Monte Carlo impurity transport simulation code (IMPMC code). The estimation of sputtering yields and development of the simulation code enabled reliable predictions for impurity behavior in a fusion reactor.
